核电厂废树脂中90Sr、55Fe含量的快速分析

随着我国核电机组不断投入运行,会产生大量的放射性废树脂,废树脂中含有的中、长半衰期放射性核素关系到处置场的长期安全性,是处置场安全评价的重点。本工作利用萃取色层法和低本底液闪测量法建立了一种核电厂废树脂中⁹⁰Sr和⁵⁵Fe快速分析方法,该方法Sr的化学回收率为94.94%,最小可探测比活度为5.1 Bq/kg,采用两种分离方法Fe的化学回收率分别为96.60%、87.23%,最小可探测比活度分别为10.83、11.76 Bq/kg。     

Moving the Largest Capacity PWR Dual-Purpose Cask in the World from Gösgen NPP to the Zwilag Interim Storage Site

Abstract

The Swiss Gösgen nuclear power plant (NPP) has decided to use two different methods for the disposal of its spent fuel. (1) To reprocess some of its spent fuel in dedicated facilities. Some of the vitrified waste from the reprocessing plant will be shipped back to Switzerland using the new COGEMA Logistics, TN81 cask. (2) To ship the other part of its spent fuel to the central interim storage facility at Zwilag (Switzerland) using a COGEMA Logistics dual-purpose TN24G cask. The TN24G is the heaviest and largest dual-purpose cask manufactured so far by COGEMA Logistics in Europe. It is intended for the transport and storage of 37 pressurised water-reactor (PWR) spent fuel assemblies. Four casks were delivered by COGEMA Logistics to Gösgen NPP. Three transports of loaded TN24G casks between Gösgen and Zwilag were successfully pelformed at the beginning of 2002 using the new COGEMA Logistics Q76 wagon specifically designed to transport heavy casks. This article describes the procedure of operations and shipments for the first TN24G casks up to storage at Zwilag. The fourth shipment of loaded TN24G was due to take place in October 2002. The TN24G cask, as part of the TN24 cask family, proved to be a very efficient solution for Kemkraftwerk Gösgen spent fuel management.     

Radiological aspects and behavior of spent fuel considering long-term storage

For spent nuclear fuel management in Germany, the concept of dry interim storage in dual purpose casks before direct disposal is applied. Current operation licenses for storage facilities have been granted for a storage time of 40 years. Due to the current delay in site selection, an extension of the storage time seems inevitable. In consideration of this issue, GRS performed burnup calculations, thermal and mechanical analyses as well as particle transport and shielding calculations for UO₂ and MOX fuels stored in a cask to investigate long-term behavior of the spent fuel related parameters and the radiological consequences. It is shown that at the beginning of the dry storage period, cladding hoop stress levels sufficient to cause hydride reorientation could be present in fuel rods with a burnup higher than 55 GWd/tHM. The long-term behavior of the cladding temperatures indicates the possibility of reaching the ductile-to-brittle transition temperature during extended storage scenarios. Surface dose rates are 3 times higher when a cask is partially loaded with 4 MOX fuel assemblies. Due to radioactive decay, long-term storage will have a positive impact on the radiological environment around the cask.     

Safety Aspects of Long-term Dry Interim Storage Of Type-b Spent Fuel and Hlw Transport Casks

Abstract

Based on the German decision to minimise transport of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities were licensed until the end of 2003. Because of the large amount of Type B(U) transport casks which are going to be used for long-term interim storage the question of timelimited Type B(U) licence maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of qualification of transport casks for interim storage is the long-term behaviour of the metallic seal–lid system. Here results are presented from current long-term experimental tests with metallic 'Helicoflex' seals in which pool water is enclosed. This series of tests has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an exchange of experience, know-how and state-of-the-art between authorities and technical experts with regard to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called 'coordinating institution for cask dispatching information' ('KOBAF') which entails management of an online database of cask-specific documents and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the coming decades.     

Mobile system for the management of fuel assemblies from research reactors

Interim storage in transport and storage casks of the CASTOR type, and later the final storage of these casks are planned for the management of spent fuel assemblies from German research reactors.A mobile transfer unit is used for loading the casks with fuel assemblies on the reactor sites. Key components of the mobile transfer unit are a transfer cask, the recharging lock, and an air-cushion transport system. By means of the air-cushion transport system, the whole equipment, as well as the CASTOR casks, is transported into the reactor building. Thus, handling of the 16 t CASTOR casks is possible even on reactor sites within sufficient crane capacity. A 20 ft container accommodates the mobile transfer unit and all accessories so that the whole equipment can be transported to the reactor sites by truck.     

废离子交换树脂的优化处理

核设施产生的废树脂的安全处理、整备和处置是热点问题。本文论述了废树脂的特殊性 ,解析了各种废树脂处理、整备技术 ,包括脱水干燥后装入高整体性容器、洗脱、热压、生物降解、焚烧、湿法氧化、沥青固化、聚合物固化、玻璃固化和水泥固化等。重点分析了废树脂水泥固化 ,讨论了树脂溶胀作用破坏固化体的机理 ,介绍了克服树脂溶胀作用的方法。强调指出必须重视水泥固化的配方 ,关键是必须满足处置要求。最后 ,对优化处理废树脂提出了建议     

Status of the FRG activities on direct disposal of spent LWR-fuel

As stipulated by the German Atomic Energy Act, reprocessing is the reference waste management route for LWR's in the Federal Republic of Germany (FRG).Spent fuel disposal without reprocessing is being developed to technical maturity for those fuel elements for which reprocessing is either technically not feasible or economically not justifiable. The reference concept for direct disposal is the emplacement of large and heavily-shielded casks in drifts of a repository mine located in a salt dome. Moreover, a back-up solution is being pursued which results in smaller canisters which are emplaced in boreholes.The mining authorities have pointed out that the feasibility of direct disposal is to be demonstrated before a license for industrial scale deployment could be granted. Demonstration tests are necessary in the following areas: shaft transport of large and heavily shielded casks, handling of the casks in the repository and thermal and rock mechanics investigations with respect to the drift emplacement concept.The results of the demonstrations tests as well as the results from layout and optimization studies for a common repository for both reprocessing waste and spent fuel will be available early enough to be incorporated into the licensing procedure for the FRG's first repository for heat-generating nuclear wastes. This means that direct disposal of spent fuel not suitable for reprocessing could be introduced in the future in addition to the reprocessing and recycling waste management concept.     

Thermal characteristics of aged spent fuel in a dry environment

Shipment of spent nuclear fuel from operating reactors is an important link in resolving the fuel storage and nuclear waste problems. Certain thermal problems must be considered. The nuclear spent fuel, even after a period of pool storage, has sufficient decay heat to necessitate special handling when being shipped to an off-site location. This paper presents the results of development related to the thermal interaction between dry spent fuel casks and nuclear fuel under operating situations. The tests were performed at the Barnwell Nuclear Fuel Plant (BNFP) using full-sized truck and rail casks and electrically heated dummy fuel assemblies. The safe and practical operation of the equipment developed has been shown.     

Heat removal verification tests using concrete casks under normal condition

Heat removal tests using two types of full-scale concrete casks were conducted. This paper describes the results under a normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced concrete cask (RC cask) and concrete filled steel cask (CFS cask) were the specimen casks. Decay heat at the initial period of storage 60 years of storage, the middle period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data requested for heat removal analyses were obtained.     

Cooldown testing of nuclear fuel casks

Shipment of spent nuclear fuel from operating reactors is an important link in resolving the fuel storage and nuclear waste problems. Certain thermal problems must be considered. The nuclear spent fuel, even after a period of pool storage, has sufficient decay heat to necessitate special handling when being shipped to an off-site location. This paper presents the results of development related to the thermal interaction between dry spent fuel casks and nuclear fuel under operating situations. The tests were performed at the Barnwell Nuclear Fuel Plant (BNFP) using full-sized truck and rail casks and electrically heated dummy fuel assemblies. The safe and practical operation of the equipment developed has been shown.     

废树脂微波处理工艺研究

废树脂微波处理工艺主要利用微波的穿透能力强和体加热的特点,首先将废树脂中的水分除去,然后将废树脂中的有机组分去除,从而达到有机废物无机化的目的。与原始废树脂相比,微波干燥、灰化处理后,得到的灰分减容和减重比均极大。利用该技术处理得到的产物,可以方便的进行后续处理。通过研究,确定了废树脂微波桶内干燥及废树脂微波灰化工艺参数。     

Analysis for seismic response of dry storage facility for spent fuel

Most of the dry storage systems for spent fuel are freestanding, which leads to stability concerns in an earthquake. In this study, as a safety check, the ABAQUS/Explicit code is adopted to analyse the seismic response of the dry storage facility planned to be installed at Nuclear Power Plant #1 (NPP1) in Taiwan. A 3D coupled finite element (FE) model was established, which consisted of a freestanding cask, a concrete pad, and underneath soils interacting with frictional contact interfaces. The scenario earthquake used in the model included an artificial earthquake compatible to the design spectrum of NPP1, and a strong ground motion modified from the time history recorded during the Chi-Chi earthquake. The results show that the freestanding cask will slide, but not tip over, during strong earthquakes. The scale of the sliding is very small and a collision between casks will not occur. In addition, the differential settlement of the foundation pad that takes place due to the weight of the casks increases the sliding potential of the casks during earthquakes.     

The economic effects of the deferred disposal of spent fuel in Korea

The present study analyzes the economic effects concerning deferred disposal of spent fuel through long-term storage. According to the cost analysis, a scenario that a 90-year deferral of an HLW (High-Level Waste) repository construction in favor of a long-term storage of spent fuel would be economically preferable to another scenario based on the year 2040 chosen as the starting point for construction on a repository. That is, the former scenario would cost about 1/2 of the latter. This finding is an estimated result from an economic perspective only, assuming the disposal of 20,000-ton PWR spent fuel and 16,000-ton CANDU spent fuel. Still, it seems necessary to elicit proper term of storage for radioactive waste in order to comply with the so-called Polluter-Pays principle that the current generation cannot pass on its radioactive waste to the next generation.     

Stabilization and volume reduction of radioactive spent ionexchange resins

Stabilization and volume reduction of spent radioactive ion-exchangeresins(IERs) were studied.Stabilization technology includes volume reduction with wet chemical oxidation process and immobilization of the residue into cement.Under suitable conditions,the exhausted radioactive ion-exchange resins were dissolved successfully in a H2O-Fe^2 /Cu^2 catalytic oxidation system(Fenton reagent).The analytical results indicated that the radioactive nuclides loaded in the resins were concentrated in decomposed solution and solid residues.The process parameters of wet chemical oxidation and solidification were also obtained.The decomposition ratios were 100% and more than 90% for cation and amion IERs respectively.The waste volume was decerased by 40% compared with that of original spent resins.     

Development of new transportable storage casks for interim dry storage

Abstract

We have started a programme to design a new type of transportable storage cask (Hitz casks) for both boiling water reactor (BWR) and pressurised water reactor (PWR) fuels for use in the new interim dry spent fuel storage project in Japan. The basic policy of this development is to use proven technology to realize a safe and cost-effective design with a high transport and storage capacity and a low fabrication cost. Since it is not permissible to change the lid gaskets at the storage facility, the double-lid system is designed to be able to use double metallic gaskets as the containment boundary for transport after the storage period; this is one of the new design features used in the casks. With the basket design we tried to achieve a capacity of 69 fuel assemblies for BWR fuel and 26 fuel assemblies for PWR fuel. Further details about these and other topics are discussed.     

Adaptation of Approved Transport and Storage Cask Lines to Individual Customer Requirements

Abstract

Within the decommissioning programmes of the Italian nuclear power plants, the Italian multi-utility company ENEL decided to rely on on-site dry storage while waiting for the availability of the national interim storage site. SOGIN (Società Gestione Impianti Nucleari SpA, Rome, Italy), now in charge of all nuclear power plant (NPP) decommissioning activities was created in the ENEL group but is now owned by the Italian government. In 2000 it ordered 30 CASTOR® casks for the storage of its spent fuel not covered by existing or future reprocessing contracts. Ten CASTOR X/A17 casks will contain the Trino pressurised water reactor (PWR) fuel and the Garigliano boiling water reactor (BWR) fuel currently stored in pools at the nuclear power plant Trino and the Avogadro nuclear facility at Saluggia. Additionally 20 CASTOR X/B52 casks will contain the BWR fuel assemblies, which are stored in the pool at the Caorso nuclear power plant. GNB (Gesellschaft fuer Nuklear-Behaelter mbH, Essen, Germany) has completed detailed studies for the design of both types of cask. The tailored cask design is based on the well-established and proven design features of CASTOR reference casks and is responsive to the needs and requirements of the Italian fuel and handling conditions. The design of the CASTOR X/A17 for up to 17 Trino PWR fuel assemblies or 17 Garigliano BWR fuel assemblies and the CASTOR X/B52 cask holding up to 52 Caorso BWR fuel assemblies is suitable for the following conditions of use: loading of the casks in the fuel pools of the nuclear installations at Trino, Caorso and Avogadro; no upgrading of the Current on-site crane capacities; transport of the fuel assemblies, which are currently stored at the Saluggia facility to the nuclear power plant Trino; on-site storage in a vertical or horizontal position with the possibility of transfer to another temporary storage or a final repository, even after a number of years; the partial loading of mixed oxide (MOX) and failed fuel; loading and drying of bottled Garigliano fuel assemblies. On the basis of the CASTOR V/19 and CASTOR V/52 cask lines, the design of the CASTOR X/A17 and X/B52 casks aims at optimising safety and economics under the given boundary conditions. The long time for which fuel is kept in intermediate wet storage results in a reduced shielding and thermal-conduction requirement. This is used to meet the tight mass and geometry restrictions while allowing for the largest cask capacity possible.     

乏燃料储存与运输两用容器安全性能研究进展

乏燃料运输和储存两用容器具备乏燃料运输和储存两种功能,是乏燃料实现最终贮存和处置前的一种储运方式。本文介绍国际乏燃料储存与运输两用容器安全设计要求和安全验证实践经验,研究适合我国乏燃料储存与运输两用容器安全设计要求和安全验证要求,为我国乏燃料储存与运输安全提供参考。     

Topics of research and development on concrete cask storage of spent nuclear fuel

This paper addresses topics of research and development (R&D) being challenged for realization of concrete cask storage of spent nuclear fuel in Japan. Comparison between metal cask storage and concrete cask storage is addressed. Background of these R&D and current status of technology on spent fuel storage are described. Need and design concepts of concrete cask storage technology, tests and evaluation of integrity of spent fuel, materials, concrete casks under normal and accident conditions, monitoring technology, etc. are systematically arranged and introduced. Topical problems of these R&D are described.     

Japanese perspectives and research on packaging,transport and storage of spent fuel

Abstract

The present paper gives an overview of Japanese experimental studies of dual-purpose metal casks. The studies included: cask drop without impact limiters, drop of a heavy weight onto a cask due to building collapse, burial of a cask in debris from building collapse, tipping over of a cask during an earthquake, long-term containment of metal gaskets and transportability of casks after long-term storage. Most of the studies employed full-scale casks for the experiments.     

Active reduction of the end effect by local installation of neutron absorbers

In the burnup credit analyses of interim or long-term spent fuel (SF) storage facilities and transport casks, when the average burnup value is greater than approximately 30 GWd/t, the neutron multiplication factor becomes greater if we consider the axial burnup distribution of the spent fuel assembly rather than assuming an average burnup. This phenomenon is called the “end effect” and it is one of the main technical issues in burnup credit research. The end effect is characterized by an increase of the neutron flux around the end regions of the spent fuel assemblies in the criticality calculation. However, such increase of the neutron flux has not been observed in experiments using actual spent fuel assemblies.